Method for measuring fissionable material content of fuels



April l, 1969 D. n.. BASDEKAS 3,436,538

METHOD FOR MEASURING FSSIONABLE MATERIAL CONTENT OF FUELS Filed Feb. l, 1966 United States Patent O 3,436,538 METHOD FOR MEASURING FISSIONABLE MATERIAL CONTENT OF FUELS Demetrius L. Basdekas, San Antonio, Tex., assigner, by

mesne assignments, to the United States of America as represented by the United States Atomic Energy Commission Filed Feb. 1, 1966, Ser. No. 523,997 Int. Cl. Glt 3/00 U.S. Cl. Z50-83.1 5 Claims ABSTRACT F THE DISCLOSURE The `present invention relates to an improved method of indirectly assaying nuclear fuel and more particularly to a non-destrutcive assay method wherein a thermal neutron beam is directed to the fuel plate under test and fission neutrons are thermalized to activate a dysprosium plate for radiation measurement whereby fuel content of the fuel plate under test may be determined.

The invention described herein was made or conceived in the course of or under a contract with the United States Atomic Energy Commission.

It is highly desirable to non-destructively measure the fissionable material content of both irradiated or hot and un-irradiated or cold nuclear `fuels with accuracy and precision. The fissionable material content of hot fuels is important in determining efficiency of reactor operations by comparing the amount of work produced to the fuel consumed as well as determining where burn-up occurs within a fuel element and `when to replace fuel elements. In addition, determination of tissionable material content of cold nuclear fuels is important in loading reactors since the total amount of issionable material and its homogeneity is critical. Of course, knowledge of tissionable material content of nuclear fuels is important for operational safety, management and accounting of fssionable materials.

Conventional methods of assaying nuclear fuel and particularly uranium include chemical analysis and gamma ray scintillation spectrometry. However, chemical analysis is difcult and costly and involves destruction of the nuclear fuel under test which is obviously undesirable in most instances. Gamma ray scintillation spectrometry has been used for the determination of the H235 and U238 content of cold reactor fuel elements as well as burn-up of hot fuels by measuring activity. However, such determination is dependent in the latter situation on the history of the fuel element operation. Operational history and activity cannot be correlated with a great degree of accuracy so it is apparent that an assay method capable of suitable accuracy would `be highly advantageous. The present invention is directed to such a method.

It is, therefore, an object of the present invention to provide an improved method of assaying nuclear fuel by non-destructive means.

A further object of the present invention is the provision of a non-destructive method of assaying nuclear fuel with a high degree of accuracy.

Another object of the present invention is the provision of an accurate method for non-destructively measuring 3,436,538 Patented Apr. l, 1969 ice the amount of U235 in cold fuel plates containing both U235 and U233' Still another object of the present invention is the provision of an accurate method for non-destructively measuring the amount of Um' burn-up in a hot fuel plate cOntaining both U235 and U2.

Yet a further object of the present invention is the provision of an accurate method for assaying uranium fuel by subjecting it to a thermal neutron beam, thermalizing the fission neutrons to activate a dysprosium plate and measuring the resulting activity.

Other and further objects, features and advantages will be apparent from the following description of a presently preferred embodiment of the invention, given for the purpose of disclosure and taken in conjunction with the accompanying drawing showing means for non-destructively assaying nuclear fuel.

The improved method of non-destructively assaying nuclear fuel in the present invention generally comprises directing a thermal neutron beam on the fuel plate under test which contains an unknown amount of U235. The thermal neutrons of the original beam transmitted through the fuel plate are eliminated by use of a cadmium plate and the fission neutrons produced in the plate are then thermalized. Such thermalized fission neutrons then activate a dysprosium plate causing decay which emits beta and gamma radiation capable of measurement. The decay activity is proportional to fuel content of the plate under test which may be determined by a proper correlation.

When U235 is bombarded with thermal neutrons, part of the neutrons are absorbed by the U235 nuclei and cause the U235 to fission. Each fission produces fission product nuclei and either two or three ssion neutrons (average 2.5 neutrons). By detecting these fission neutrons, the number of tissions and hence the amount of U235 can be determined such as by the method of the present invention as generally shown in the drawing. Of course, if the U235 is contained in a mixture of U235 and other elements, allowance must be made for the absorption of thermal neutrons by the other elements.

With reference to the drawing, the fuel plate under test is for example a typical mixture of U235 and U2. The thermalized neutron beam may be emitted by a neutron generator having as an example an output of 25x10 fast neutrons (14 mev.) per second `when the generator is operated at a 2.5 milliampere beam current at kilovolts. These fast neutrons can be thermalized by suitable means such as placing the target end of the generator in the center of a cubicle of water so that the resulting beam is on the order of 5 lt)B to 109 thermal neutrons/cm?! second.

If the fuel plate `under test contains U235 and U23", both U235 and U238 will absorb the thermal neutrons in the original beam. The U235 will either become U23E or will fission. The number of absorbed thermal neutrons which cause the U235 to fission can be determined by comparing the fission cross-section of the U235 to the total absorption cross-section of U235 and U238 mixture as will be explained hereafter.

As shown in the drawing, the thermal neutrons of the original `beam transmitted through the fuel plate under test are eliminated by the use of a suitable cadmium plate. Other elements having high neutron capture cross-sections may also be used to reduce the number of neutrons in the original thermalized neutron beam passing through the fuel plate under test.

The fission neutrons from the fuel plate being inspected may be thermalized such as by heavy water (D20) or water (H2O). Heavy water is preferable for the moderating material because its moderation characteristics provide a good compromise between neutron absorption and loss of neutrons because of geometry. Both heavy water and water have low neutron absorption cross-section and large quantities are not required for neutron thermalization. If large quantities of a moderating material were required, the loss of neutrons would be large because of the inverse square relation between distance and ux.

After thermalization, the fission neutrons are allowed to strike a dysprosium plate causing activation of Dyl" which becomes Dy165m with emission of beta and gamma radiation which may be detected by suitable means. For example, gamma radiation may be detected and counted by a scintillation detector. The number of gammas counted in a specified length of time is proportional to the incident fiux of thermalized fission neutrons which in turn is proportional to the U235 content of the fuel plate under test.

Alternatively, a detection method which provides a permanent record of the information extracted is autoradiography whereby fuel content of the plate under test may be determined from lm density after the film has been developed under standard conditions. A calibration curve of the film density versus fuel content may be established experimentally.

If the fuel under test has been used in a reactor, there will be U235 fission, products, Prim, Pu239 fission products and U236 in the fuel in addition to the U235and U2. The Pu239 is produced when a U233 nucleus absorbs a neutron, becomes radioactive, and decays to Np239, which decays to Pum. The Pu239 nuclei may then absorb neutron and fission, producing Pu239 fission products and neutrons. As these other products are produced by the U235 and U23, the amount of U235 and U23B decreases. The percent decrease of U235 and U23E in a reactor depends on the absorption cross-sections of U235 and U238 and the power produced by the reactor. The decrease of 11235 and U23B content and production of other elements must be considered when the U235 content of hot (reactor irradiated) fuel plates is determined.

To illustrate the present invention in nuclear fuel assay, the following examples indicate through calculations the application of the method of the invention. The fuel plates under test are assumed to be 24S/s inches by 2.8 inches by 0.060 inch wherein of the 0.060 inch dimension cladding thickness is assumed to be 0.010 inch total. The percent burnup is assumed to be the percent loss of U235 as compared to the original amount of U235 in the fuel plate.

Example I Assume that a 90.00% enriched, cold fuel plate is to be assayed by use of a neutron generator operating at 150 kv. potential and 2.5 ma. deuteron beam current having an output of 2.5 l011 neutrons/sec. over 41r geometry. Thermalization of these neutrons by a 4-foot cubicle of water will produce a thermal neutron flux of 5 10B to 109 thermal neutrons/cm-2/sec.

If 5 l0E thermal neutrons/cm.2/sec. are incident on a 2.8 inch 2.8 inch (50 cm?) area of the fuel plate, the total number of neutrons per second striking the plate is approximately 2.5 101.

Of the 2.5 l010 thermal neutrons/sec. striking the plate the number which will `be absorbed in the plate is Nor-number of thermal neutrons incident on the plate per second N=number of thermal neutrons absorbed in the plate per second Ef=total macroscopic absorption cross-section of the fuel (crrL-l) tzthickness of the fuel in the plate (0.050 in.=0.127

The absorption of the cladding is neglected since its thermal neutron absorption cross-section is considerably less than that of the uranium.

The macroscopic thermal neutron absorption crosssection of the fuel is Eafza2s5+2azas where E235=macroscopic thermal neutron absorption crosssection of U235 in the fuel Eamzmacroscopic thermal neutron absorption crosssection of U23 in the fuel The macroscopic thermal neutron absorption crosssection is related to the microscopic cross-section by the equation where N'=number of atoms of the material present Ua=microscopic thermal neutron absorption cross-section.

Thus for 90.00% enriched fuel where NA=Avogadros number (atoms/mole) puzdensity of uranium (grams/cm3) 235=atomic weight of U235 (grams/mole) Likewise for 90.00% enriched fuel N Ps Zazazgaass Qg Substituting the numerical values for 0,235, fam, NA, and au 2,335: (092 l0 24 mu/nucleus) (0.0221l l023 atoms/mole) (18.8 g./cui.3)(0.9000) 235 g./mole 2am: (2.8 X l0*24 Gru/nucleus X (0.023 l023 ntoms/mole)(18.8 g./cm.3)(0.l000`) 238 g./mole :0.0l33 cm.-x

and

Z.f=30.0090 exhibi-0.0133 cm.1=30.0223 cnn.rl

The number of thermal neutrons absorbed per second in the fuel plate is then N =Nn(1- e-a't) N: (2,5 X 1010) (1 e-3o.ozza a.121) N=2-4448 101D thermal neutrons/sec.

rq=fission neutrons produced per thermal neutron absorbed i/:fission neutrons produced per fission Eafzmacroscopic thermal neutron absorption cross-scction of the fuel Zff=macroscopic thermal neutron fission cross-section of the fuel Since the thermal neutron fission cross-section of U23H is less than 0.5 millibarns, the thermal neutron fission cross-section of the fuel is essentially equal to the thermal neutron fission cross-section of U235 in the fuel.

The thermal neutron fission cross-section of U235 in 90.00% enriched fuel is faenas?? (0.0000) (6.023X1023 atoms/molo) (18.8 g./cm.3) (0.9000) X 235 g./rnole 25.152 em.-l

=2.095 fission neutrons produced/absorption Then Nr: (2.4443 1010) (2.095) Nt=5.l2l8 X 101D fission neutrons produced/sec.

To eliminate the flux of thermal neutrons transmitted through the fuel plate, a cadmium sheet is inserted in the thermal beam between the fuel plate and the detector. 1f the cadmium sheet is 0.030 inch thick, the thermal neutron flux is reduced by a factor of 104, but the fission neutron ux will be reduced by only 2% For simplification of calculations, assume that the fis- Sion neutrons are emitted by a plane disk of 50 cm.2 area (radius=4 cm.) rather than a square plate of 50 cm.2 area (2.8 incheszll cm. on a side). The fission neutron ux at a point on the axis of a disk is approximately the same as the fission neutron ux at a point on the axis of a plate. n

The equation of the uX on the axis of a disk is where SA=source strength (neutrons/cm-2/sec.) R0=radius of the disk (cm.) a=distance from disk to point of measurement (cm.)

If the distance from the fuel plate to a dysprosium plate is ll centimeters, the fission neutron flux at the plate would be 5.1218Xl0m neutrons/sce.

e: (2.5609 S) (0.1242) d1=3.l806X107 neutrons/cm.2/sec.

The total number of ssion neutrons per second striking the dysprosium plate would be approximately equal to the neutron flux at the center of the dysprosium plate times the area of the dysprosium plate. Hence, the total number of fission neutrons striking a plate having an area of 50 cm.2 would be I A=(3.l806 X107 neutrons/cm-2/sec.) (50 cm?) QA: 1.5903 X 109 neutrons/sec.

In the above calculations fission neutrons have been considered as striking the dysprosium plate. In actuality, the fast fission neutrons will cause very little activation of the dysprosium. The region between the fuel plate and the dysprosium plate must contain a moderating n1aterial. In this ease the moderator considered is heavy water, D20. The thermalization distance for fission neutrons in heavy water is ll centimeters, which was used in the previous calculations.

Additional heavy water may be used around the path from the fuel plate to the dysprosium plate. This heavy water will act as a reflector to contain some of the thermalized neutrons which otherwise would diffuse from the beam.

Because the loss of fission neutrons from the beam is practically inestimatable generous allowance for this loss must be made with an assumption that the loss of neutrons from the beam reduces the neutron flux by a factor of 10.

If the thermal neutrons are allowed to strike a dysprosium plate, the production of radioactive Dym nuclei is given by (1N, I I0( 1 f2-:Nylon where dNtznumber of Dym nuclei produced per increment of time (dt) lU-:number of thermal neutrons incident per second EaDyrmacroscopic thermal neutron activation crosssection of Dym for Dy165m production (cml) dzdysprosium plate thickness (cm.)

dt=incremental exposure time The decay of radioactive nuclei during the exposure of the dysprosium plate is given by (JAH- dt.

)Nvt n where Ntztotal number of Dy165m nuclei present at any time (t) ^Dy=decay constant of Dy165m=0-00924 sec.1

Combining the concurrent rates of activation and decay and integrating to find the number of radioactive Dy165 nuclei present after an exposure time (f) where Nznumber of radioactive Dylsm nuclei present after exposure time (v) fzexposure time where aaDY=microscopic thermal neutron activation cross-section of Dy64 NA=Avogadro`s number (atoms/mole) pnydcnsity of dysprosium 7 164=atomic Weight of the activated target nuclei, dysprosium -164 0.282:abundance of dysprosium -164 in natural dysprosium Hence zam: (2600X 1024 cm.2/1n1c1e\1s) 2a a o (6.023x 10 mole (8.50 gJm. )(0.28.)

104 g.,fmol0 23.0 crnl The decay constant for Dy165m is loge DY: ,lvl/2D),

where TU2DY=halflifc of Dylfi5m=75 sce.

Then

Dy 75 Sec' 0.00924 scc.

The number of radioactive Dy165m nuclei present after exposing a 0.02 centimeter thick dysprosinm plate having an area of 50 cm.2 for 20 seconds is then 1*.6003X 10B 0.00944 Nf:(1.7211 1010)(1-0.632)(1-0.832) Nr=(1.7211 1010) (0.368) (0.168) Nr:l.645 10g radioactive nuclei If the activated dysprosinm plate is counted for one half-life (75 seconds), the total number of disintegrations counted for 21r geometry is D=NrCG 1.0645 109 nuclei) (1/2 decaying) (2r/411- geometry) =2.6612 108 counts where Dznumber of counts detected C:fraction decaying Gzgeometry factor It has been assumed that the detector is 100% efficient, which is essentially correct for scintillation detection of the low energy gamma radiation (108 kev.) emitted by Dy165m.

The standard counting deviation for 2.6612 108 counts to 95% confidence is NT: (1 -6-2311X0-02) (1 e-D-00024X20) Percent Standard Deviation: i5-|- 100% If the activated dysprosinm plate is counted for 20 seconds, the total number of nuclei decaying in this time is 1.7884Xa nuclei. Thus gamrnas will be emitted over 21r geometry, and may be counted by a detector.

The standard deviation in this case is Percent, Standard Deviation:

3 vmxw lf the enrichment of the fuel plate were 90.02%, the

activation would change by approximately 0.02/9000 or 0.022%.

Since this change in activation is larger than the standard deviation for the technique using a -second counting time, a difference between 90.00% enrichment and 90.02% enrichment should be detectable in cold fuel plates.

Example II Assume now that a 1.140% enriched cold fuel plate is to be assayed as in Example I.

The macroscopic thermal neutron absorption cross-section of U235 in a 1.140% enriched fuel plate is NAPlx 2,235 @mei-3 (0.01140) 0.3801 cm.-l

The macroscopic thermal neutron absorption cross-section of U238 in a 1.140% enriched fuel plate is Z 235:., zasNApB :0.1317 ern.1

The macroscopic thermal neutron `absorption cross-section of the fuel is then 2er: 25235'1- :B238

: (021801+01317) enr-1:05118 om.l

The macroscopic thermal neutron fission cross-section of the fuel is The number of fission neutrons produced per thermal neutron absorbed in the fuel is :1.5563 fission neutrons/ thermal neutron absorbed in the fuel The number of thermal neutrons per second absorbed in the fuel is :1.5730 10 thermal neutrons/scc.

The number of fission neutrons produced per second is then Nx:N11

:24481)(10q ssion neutrons/sec.

The ssion neutron fiux at the dysprosium plate would be RoLi-Ul2 2.4481 X l0 neutronfsec.

50 cm.2 lo 4 ge Allowing the thermal neutrons to strike the dysprosium plate, the number of radioactive Dy*65m nuclei produced in 20 seconds is 105 -23 0X0 02 0.00924X20 0.09324 (1 e )(1 e :5.0860X107 radioactive nuclei For 75 seconds counting, the total number of counts detected is D=NCG D=(5.0860 107 nuclei)(1/2 decaying)(2vr/41r geometry) :1.2715 107 counts The standard counting deviation for 95% confidence is For 20 seconds counting, `8.5445 106 nuclei will decay, and (8.5445 106) (Z1r/41r)=4.2722 l06 counts will be detected for 21r detecting geometry.

The standard deviation in this case is Percent Standard Deviation: X 100% Percent Standard Deviation= X 100% If the enrichment of the plate were to be changed from 1.140% to 1.141%, the number of counts detected should change by approximately 0.00l/l.l4l or 0.09%. Thus a change from 1.140% to 1.141% enrichment should be detectable above the standard deviation of y0.084396.

Example III Assume now that a 90.00% enriched plate having 2% burn-up is to be assayed. In hot fuel assay, the production of U235 fission products, plutonium-239, U236, and P0239 fission products must be considered.

The macroscopic thermal neutron absorption crosssection of U235 in a 90.00% enriched fuel plate after 2% burn-up is:

The macroscopic thermal neutron absorption cross- Section of U2118 in a 90.00% enriched fuel plate after 2% burn-up is The macroscopic thermal neutron absorption crosssection of U235 fission products in a 90.00% enriched fuel plate after 2% burn-up is :0.3235 crnfl The production of Pu 239 is related to the burn-up of U2x5 by the ratios of the cross-sections and abundances of U235 and Um. The relationship is (67013233) (atm)A (930235) (am) (0.1 (2.8)

:0.00001075 of the original U255 Since the thermal neutron absorption cross-section of Pu239 is of the same order of magnitude as that of U235 (0,1m: 1060 barns) Prim produced=U235 burn-up may be U25*6 produced: (U235 burn-up) Where cr235=mieroscopic thermal neutron capture cross-section of U2315 Hence 112 barns 580 barns :0.00386 of original U235 Since the thermal neutron absorption cross-section of U236 is l6 barns compared to 692 barns for D235, the effect of absorption by U235 may be neglected.

The thermal neutron absorption cross-section of 90.00% enriched fuel after 2% burn-up is :at 2333+ 2.233 zur.

: (29.4088 -i- 0.0133 -l- 0.3235) cmil :29.7456 cm.-l

The thermal neutron fission cross-section of the 90.00% enriched fuel after 2% burn-up is Eff: ifmwa (0.0000) (0.03)

the U236 produced: (0.02)

:24.649 cm.-l

The number of fission neutrons produced per thermal neutron absorbed in the fuel is 27 f f (2 5) 24.649 om.n1 2g* 20.7456 cnt-1 :2.072 fission neutrons producedfabsorption Assuming a thermal neutron fiux of 2.5 X 101 thermal neutrons/cm.2/sec. is incident on the fuel plate, the number of thermal neutrons per second absorbed in the fuel is N:N( 1 -e-Ef) (2 5X1010)(1 e-zmmoxoaz) :2.4428X 10lo neutrons/sec.

The number of fission neutrons per second produced in the fuel is :5.0615X 10I fission neutrons/sec.

The fission neutron fiux striking the dysprosium plate would be 50615)( 1010 neutrons/sec.

50 cm.s toga [(4 cm.)2+(11 cm.)]

4 (l1 cm.)z

: 3.1432 X 107 neutrons/cm/sec.

The total number of fission neutrons per second striking the dysprosium plate would be I A=(3.l432 10'z nts./cm.2/sec.) (50 cm?) :1.5716 X109 neutrons/sec.

For 20 seconds irradiation of the dysprosium plate, the number of radioactive Dy165m nuclei produced will be :1.0510X g radioactive nuclei For 75 seconds counting the total number of counts detected over 21r geometry is D=N,CG D: 1.0510 109 nuclei) (1/2 decaying )(21r/'41r geometry) :2.6275 X103 counts The standard counting deviation to 95% confidence is Percent Standard Deviation:

j 0., z C w/2.6275 10BX100/0 0.0185

Percent Standard Deviation: 100 72;

A 90% enriched fuel plate will produce a count of 2.6612 10 counts in a 75-second decay period before it experiences any burn-up. After irradiation to 2% burnup the total count drops to 2.6275X103 counts in a 75- second decay period. The change in count for a 2% burn-up is then For 0.03% burn-up the decrease in total count should be approximately 0.019%. Hence a burn-up of 0.03% should be detectable above a 75-second count standard deviation of 0.0l851%.

Example IV Assume that a 1.140% enriched plate having 2% burn-up is to be assayed. Again the production of U235 fission products, plutonium-239, U23, and Pu239 ssion products must be considered.

The macroscopic thermal neutron absorption crosssection for U235 in a 1.140% enriched fuel plate after 2% build-up is `',IVAP

:0.3725 em.-l

The macroscopic thermal neutron absorption crosssection for U235 fission products in a 1.140% enriched fue] plate after 2% burn-up is Etf-D: (0.0111292 =(0.0i1)(0.3725) cnr-1 :0.0041 @nt-1 The amount of 12u23 produced in a 1.140% enriched fuel plate after 2% burn-up is Prim Production: (002 bum-up) The production of Pu23 related to the original amount of 1.1238 is original U235 original U232 :0.0000965 of original U23B 12u23 Production (0.008373 Since the total plutonium-239 produced is only a small part of the original amount of D235 and since only a small part of the Pu23B produced will fission (only 2% U235 fissioned and its fission cross-section is comparable to that of PU239), the effect of Pu239 fission products may be neglected.

The efect of absorption by U23 may also be neglected Since its quantity is only 0.386% that of U235 and its absorption crosssection is 6 barns compared to 692 barns for U235.

The macroscopic thermal neutron absorption crosssection for Pu239 in a 1.140% enriched fuel plate after 2% burn-up is NMu 30: 219i. ...d :B2 (rn :(1000X10-24 em.2/nuclei) X (0.023 1023 atoms/mole) (18.8 g./cm.3)(0.98860)(0.0000965) 238 g./mole :0.00481 ctn-1 NAp'lL The macroscopic thermal neutron absorption crosssection for the fuel is then zf.= 2235+ :w+ 2228+ 2u- (0.3728+0.0048+ 0.1317-l- 0.0041) emr The macroscopic thermal neutron fission cross-section of the fuel is the sum of the fission cross-section for U235 and Pu239.

ztf=zf25+zfm The macroscopic thermal neutron ssion cross-section for U235 in a 1.140% enriched fuel plate after 2% burnup is NAP 235 :0.3122 cm.-l

zrzsszmzas (0.0l)(0.98)

The macroscopic thermal neutron ssion cross-section l 3 for Pu'-'39 in a 1.140% enriched fuel plate after 2% burnup is NAPu 238 (750 X 24 curl/nuclei) X (6.023 X 1023 atoms/mole) (18.8 gvemmoessemtdocooge) 238 g./mole 2 im: Urea :1.5377 fission neutrons produced/absorption The number of thermal neutrons per second absorbed in the fuel is N=Nt1-e2.l)

=(2 5X101n)(1 e-o.51a1 tt.12v) =1.57685X10 thermal neutrons/sec.

The number of fission neutrons produced per second in the fuel is NfzNn: (1.57685 X 109) 1.5377) The fission neutron ux striking the dysprosium plate would be S R 2 2 tbzloge( D Vacuum :5.0374X107 radioactive nuclei For 75 seconds counting, the total number of counts detected over 27T geometry is D:N,CG

= (5.0374 10"' nuclei) (1/2 decaying) (21r/41r geometry) 1.25935 X107 counts 14 The standard counting deviation to confidence is :0.0845570 For 20 seconds counting, 8.4628 106 nuclei will decay, and (8.4628 10B)(21r/41r)=4.23l4 106 counts will be detected over 2r geometry.

The standard counting deviation in this case is :014.50% A 1.140% enriched unirradiated fuel plate will produce a dysprosium-165m count of 1.2714X l0"I counts for 75 seconds Dy165m decay. After irradiation to 2% burn-up the total count drops to l.25935 l0"I counts for 75 seconds decay. The change in count for 2% burn-up is then Percent Standard Diviation: X

Percent, Standard Deviation 100 For 0.20% burn-up the decrease in total count should be approximately 0.0954%. Hence, for a 75-second counting period a burn-up of approximately 0.20% should be detectable above the standard deviation of 0.084596.

In operation, and with reference to the drawing, and the examples, the fuel plate under test is subjected to a thermalized neutron beam whereby fission of U235 in the fuel plate occurs. A cadmium plate or shield is provided to lter or contain the original thermalized neutron beam while the fission neutrons are thermalized preferably by heavy water, although water is also useful as a moderator. The thermalized fission neutrons then strike a dysprosium plate of optimum thickness whereby the fission neutrons are captured resulting in activation of the dysprosium. The beta-gamma activity is then measured and correlated to determine the content of the fuel plate under test. Thus provided is an accurate, non-destructive method of assaying nuclear fuel.

The present invention, therefore, is well adapted to carry out the objects and attain thc ends and advantages mentioned as well as others inherent therein. While a presently preferred embodiment of the invention has been given for the purpose of disclosure, numerous changes in the details of construction and the combination, shape and size and arrangement of parts may be resorted to without departing from the spirit and scope of the invention as hereinafter claimed.

What is claimed is: l. An improved method of non-destructively assaying nuclear fuel containing U235 including,

directing a thermal neutron beam to the fuel under test whereby fission neutrons are produced in proportion to the amount of Uim5 present in the fuel,

eliminating thermal neutrons of the original beam transmitted through the fuel under test,

thermalizing the fission neutrons,

activating a dysprosium plate with the thermalized ssion neutrons, and

measuring the activity of the dysprosium plate.

2. The invention of claim 1 wherein the thermal neutrons of the original beam transmitted through the fuel under test are eliminated by a cadmium shield.

3. The invention of claim 1 wherein the fission neutrons are thermalized by a moderator selected from the group consisting of heavy Water and water.

4. The invention of claim 1 wherein activity of the dysprosium plate is measured by gamma scintillation detection.

5. The invention of claim 1 wherein thermal neutrons of the original beam transmitted through the fuel under test are eliminated by a cadmium shield,

the ssion neutrons are thermalized by a moderator se- 1 5 1 6 lected from the group consisting of heavy water and 3,141,092 7/1964 Weinberg Z50- 83.1 X water, and 3,222,521 12/1965 Einfeld Z50-83.1

activity of the dysprosium plate is measured by gamma scintillation detection, RALPH G. NILSON, Primary Examiner.

References Cited 5 A. B. CROFT, Assistant Examiner.

UNITED STATES PATENTS Usl CLXR.

2,845,544 7/1958 Seaborg et al Z50-83.1 250 83 2,971,094 2/1961 Title Z50-83.1 

